Many electric power generating stations employ a nuclear boiling water reactor for heating feedwater to generate steam. Typically, the steam is piped from the nuclear boiling water reactor to a turbine-generator where it is used to produce electric power. The spent steam is piped from the turbine to a condenser to generate condensate. For most conventional systems, the condensate is eventually returned to the nuclear boiling water reactor in the form of feedwater by a feedwater injection system. The feedwater is eventually re-heated by the reactor core to form steam, thereby completing the steam generation cycle.
Fuel assemblies in the reactor core heat the returned feedwater to generate a two-phase steam/water mixture. The two-phase steam/water mixture naturally rises from the core to the core shroud head. The core shroud head overlies the core and defines an upper core plenum region ("upper plenum region").
In many conventional reactors, this steam/water mixture is not produced uniformly throughout the reactor core. The upper plenum region is employed to help the steam and water form a more homogenous mix.
Standpipes exit from the top of the core shroud head or upper plenum region and convey the two-phase steam/water mixture from the upper plenum region to a steam separator assembly. The steam separator assembly classifies the steam from water in the two-phase mixture. The steam is piped to the turbine-generator. The separated water is returned to the reactor vessel and is used as a coolant for the reactor core.
The primary reactor coolant inventory consists of two components. The returned feedwater makes up part of the primary reactor coolant inventory. The separated saturated water, makes up the remainder of the primary water coolant inventory for many conventional boiling water reactors.
The separated water mixes with the return feedwater in a region near the wall of the reactor vessel called the downcomer annulus region.
This downcomer annulus region transports the coolant inventory from the top of the reactor to the bottom of the reactor along a flow path outside of the core but inside of the reactor vessel walls.
The feedwater and separated water mixture thus form the primary water coolant inventory for the reactor. During normal operation of the reactor. this primary water coolant is maintained at a normal water level above the core shroud head. The primary coolant is subject to natural circulation as it migrates down the downcomer region. Coolant circulation may be assisted by the use of forced circulation pumps.
During normal operation of the reactor, the primary coolant inventory is maintained at very high temperatures and pressures. For a boiling water reactor. the vast majority of coolant is comprised of saturated water at 546.degree. F. and 1020 psia.
Various postulated events may cause the reactor to experience a partial loss of its coolant inventory. A subcategory of such loss-of-coolant inventory events includes loss-of-coolant accident (LOCA) events. in which an hypothesized pipe break results in the reactor coolant inventory to be expelled from the reactor due to the initial high pressure and temperature of the coolant. Isolation valves are installed on lines connecting to the reactor to prevent or at least mitigate the extent of coolant inventory loss. However, in certain other hypothetical situations, such as for pipe breaks occurring between the reactor and the pipeline innermost isolation valves, the action of isolation valve closures could not prevent the reactor from undergoing a full blowdown. Additionally, event scenarios may include an intentional reactor full controlled depressurization because of the hypothesized non-functioning of high-pressure coolant injection systems and the consequent need to depressurize promptly so that low-pressure coolant injection systems can accomplish needed coolant resupply. To prevent serious core damage given such casualty scenarios, it is necessary to design the reactor to have enough initial coolant inventory to keep the nuclear core at all times covered with coolant. This prevents the fuel rods of the nuclear fuel assemblies from heating beyond acceptable levels.
Another subcategory of loss-of-coolant inventory events includes events featuring a loss of all power to the station, in which case the reactor can be deprived of feedwater injection.
As a first mitiating action in a loss-of-coolant accident, the plant is automatically scrammed--i.e., all control rods are immediately inserted in the core to shut down the nuclear reaction. However, the reactor core will continue to generate significant quantities of decay heat even after the control rods have been inserted. Thus, some form of emergency core cooling is provided on conventional nuclear boiling water reactors to keep the fuel rods of the nuclear fuel assemblies from heating beyond acceptable levels.
Several emergency core cooling schemes have evolved to insure that the reactor core is properly cooled during a LOCA. The complement of safety grade systems that are provided as part of the nuclear steam supply system to meet these needs for adequate assured core cooling, are known as the emergency core cooling system (ECCS). For example, one BWR product line produced by GE Nuclear Energy uses both high-pressure as well as low-pressure injection of water as major elements comprising its ECCS. Considerable energy must be expended during the casualty to effect the required high-pressure injection of coolant. ECCS systems providing these high pressures must be brought on line while the casualty is occurring. ln addition, these systems depend on the long term operation of power supplies such as emergency diesel generators and connected electrical pumps and are therefore expensive when designed to the required margins of reliability.
Advanced designs such as the simplified boiling water reactor (SBWR) seek to avoid reliance on pumping systems during a LOCA. These systems employ in new ways the large pool of water known as a "suppression pool" which is connected by pipes to the reactor. The suppression pool in the SBWR design is located within the reactor containment at an elevation higher than the core. The water in the suppression pool can now be used to flood the reactor core by gravity action alone after the reactor has depressurized following a LOCA.
The SBWR ECCS now consists of the aforesaid suppression pool, plus the aforesaid injection lines which connect the suppression pool to the reactor, plus depressurization subsystem. Several limitations arise in connection with this conventional SBWR ECCS. First, the depressurization subsystem is required to reduce reactor pressure very rapidly. In addition, an adequate initial inventory of reactor coolant must be contained within the reactor vessel to counterbalance the coolant inventory lost because of flashing during depressurization. The initial water inventory must be such that the residual water inventory after flashing will keep the core covered by coolant until additional water is gravity injected by the suppression pool. As a result, the SBWR is required to have more initial water inventory than is needed by a conventional BWR. Approximately 15 to 20 feet of extra reactor vessel height is required to meet the needs for emergency core cooling for the SBWR.
For SBWR, this 15- to 20-foot region is also used as a chimney that promotes coolant circulation through the fuel assemblies. The two-phase steam/water mixture generated by heating water in the reactor core naturally up-flows through this region from the reactor core through standpipes to a steam/water separator assembly. Saturated liquid separated from the two-phase mixture by the steam separator assembly is discharged back into the reactor region external to the chimney. The discharged saturated water then flows at low velocities back into the reactor downcomer where it undergoes mixing with the cooler feedwater being returned to the reactor. The now-mixed coolant is at reactor pressure and is 20.degree. to 30.degree. subcooled, and so is still extremely hot. Because of its high temperature, a substantial fraction of this "hot" coolant will flash into steam during reactor depressurization following a LOCA.
For reactor coolant initially at 546.degree. F. and 1020 psia, depressurization of the reactor to 212.degree. F. and atmospheric pressure following a LOCA will result in approximately one-third of the water mass being flashed as steam and two-thirds remaining as water. The additional reactor vessel height and volume required to compensate for this coolant inventory lost due to flashing during depressurization of the reactor leads to a substantial increase in the capital cost of the nuclear island portion of the power station.